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Construction codes developed for prototype FBR Monju
Nuclear Engineering and Design 98 (1987) 283-288 North-Holland, Amsterdam
CONSTRUCTION
283
CODES DEVELOPED FOR PROTOTYPE
FBR MONJU
K u n i h i r o I I D A 1, Yasuhide A S A D A 1, K u n i o O K A B A Y A S H I 2 a n d Takashi N A G A T A 2 l Facul(vof Engineering, The Universi O' of ToA3o, 7-3-1 ttongo, Bunkyo-ku, Tokyo ll3, Japan -~Power Reactor and Nuclear Fuel Development Corporation, Sankaido Bldg. 1-9-13 Akasaka, Minato-ku. Tokyo 107, Japan
Received May 1986
The paper describes the regulation system for LWRs in Japan and discusses some typical points which are attained in the newly developed CCPBR (Construction Code for Prototype Breeder Reactor). Revised welding standards are briefly introduced for the Monju construction.
1. Introduction The safety evaluation has been accomplished in May 1983 with Japanese Prototype Breeder Reactor "Monju" which has been established after around two decades research and development activities at Power Reactor and Nuclear Fuel Development Corporation (PNC). Following this evaluation, Monju is now proceeding with an application for Construction permit to the government. The Japanese Government has been preparing necessary codes, standards and guides for regulating construction and operations of nuclear power plant stations. However, these codes/standards/guides are, almost in all the cases, applicable to Light Water Reactors. Some modifications or revisions are necessary for their application to Monju. The main part of these revisions is in the area of construction, examination and testing of structures/ components and welding of Monju. For this purpose, PNC has proposed "Elevated Temperature Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor" ("ETSDG") which enables design of structures/components in creep range with some additional considerations taking into account some other characteristic features of LMFBR structures, that is, being subjected to high/frequent thermal transients, use of sodium coolant, and thin walled structure. The Construction Code for Prototype Breeder Reactor (CCPBR) has been developed in 1983-84 to give guidance for structural design, examinations and testing of Monju on a basis of ETSDG. The present paper introduces, at first, the regulation system for LWRs in
Japan and then discusses some typical points which are attained in the newly developed CCPBR. Revised welding standards are also briefly introduced for the Monju construction.
2. Current regulation system for LWRs in Japan Agency of National Resouces and Energy (ANRE) of the Ministry of International Trade and Industries (MITI) is the responsible organization in Japan for construction and operation of nuclear power stations in Japan. ANRE has prepared codes, standards and guides for regulating the structural design, examination and testing for construction and operation of nuclear power plants. The regulation is principally based upon Electric Utility Industry Law (EUIL) which gives three articles with respect to structural design of nuclear power plants, that is, Art. 41 (Construction Permission), Art. 46 (Examination of Welding) and Art. 48 (Maintenance). Some ordinances are prepared on a basis of these articles. MITI Ordinance No. 62 (1965) "Technical Standard for Facilities of Nuclear Power Plant" gives Art. 5 (Resistance against Earthquake), Art. 9 (Materials and Structural Design), Art. 10 (Safety Valves), Art. 11 (Pressure Test) and Art. 12 (Surveillance Test). Detailed descriptions of these articles are given in Notice and Regulartory Guide. MITI Ordinance No. 81 gives "Technical Standard for Weldment of Structures for Electricity". ANRE Ordinance No. 9683 (1975) gives "Notification as to Licensing of Welding Method". Details regarding structural design, examinations and
0 0 2 9 - 5 4 9 3 / 8 7 / $ 0 3 . 5 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics Publishing Division)
284
K. hda et al. / Construction codes fi~r Prototype FBR Mo±uu
testing are given in MITI Notice No. 501 (1980) "Technical Standard for Structural Design of Facilities of Nuclear Power Stations" which is constructed upon a same concept as that of A S M E Boiler & Pressure Vessel Code Section III, although some revisions are included in order to adjust the descriptions to Japanese social circumstances. An aseismic design code is given by MITI Regulatory Guide "Standard for Design of
l
[
Nuclear Power Plants against Earthquake". m detail. The relation between these codes, standards and guides are schematically given in fig. I with regard It, articles relating structural design, examinations and testing. An important restriction m development of C C P B R is to incorporate ETSDG with current regulation systems by conserving the current c o d e / s t a n d a r d system for LWRs.
ELECTRIC UTILITY INDIJSTRY[,AN
r
!
&rt.41
Construction Permission
Art.46
Examination of Weldment
~ Art.48
]
Maintenance
MITI ORDINANCENO. 62 (1965) Technical Standard for Facilities of Nuclear
MITI: Ministry of I n terna t i ona 1 Trade
Power Plant Art. 5 Resistance against Earthquake
and Industry
Mr. 9 Material and Structural Design Art. lO Safety Valves
E
Art.t1
Pressure Test
iI
Art.12
Surveillance Test I
MITI NOTICENO. 501 (1980)
J
Technical Standard for Structural Design, etc. of Facilities of Nuclear Power Plant MITI REGULATORYGUIOE Standard for Design against Earthquake of Nuclear Power Plant MITI ORDINANCENO. 81 (1970) Technical Standard for Weldment of Structures for Electricity ANRE NOTICENO. 9683 (1975) Notification as to Licensing of Welding Methods
ANRE: Agency of Natural Resources and Energy
Fig. 1. Construction codes for LWRs under the Electric Utility Industry Law.
K. lida et al. / Construction codes for Proto(vpe FBR Monju
3. Construction Code for Prototype Breeder Reactor (CCPBR) In this section, topics of C C P B R are briefly introduced with their fundamental concepts a n d / o r backgrounds. C C P B R resulted from an incorporation of MITI Notice No. 501 with E T S D G which has been developed at PNC after around two decade research activities. E T S D G will be introduced in this seminar by the same authors as this one. E T S D G is established upon a same basic concept to that of A S M E Code Case N-47. However, E T S D G includes some advanced procedures regarding creep-fatigue evaluation, elastic follow-up evaluation and others on a basis of elastic analyses. Fig. 2 shows a conceptual framework of CCPBR.
3. I. Classification o f components In CCPBR, a classification of c o m p o n e n t s is at first denoted. This classification is made principally upon a degree of importance of the role of c o m p o n e n t s in plant safety design. Therefore the classification of components in Monju is a little bit different from that in L W R s according to the difference of system safety of L M F B R and LWR.
285
A first revision of classification is made with some concepts in a primary coolant loop. A shield plug of a reactor vessel, a cover gas system and a overflow system in a primary coolant loop are categorized as Class 3. On the other hand in LWRs, most of c o m p o n e n t s belonging to the primary loop are categorized as Class 1. This revision is made on a base of a heat removal from the core of L M F B R in which pressurization of a coolant (sodium) is not necessary for this purpose but a necessary point at this viewpoint is to maintain a level of a coolant in the core. That is, a function of the heat removal from the core is satisfied as long as a level of the coolant is kept and is not affected from a depressurization of it, in case of LMFBR. Please note that Class 3 means in Japanese system a class corresponding to Class 2 in A S M E Code. Class 2 in Japanese system is used for categorizing containment vessels which belong to Class MC in A S M E Code. Another revision of classification is made with all c o m p o n e n t s other than Class 1 which contain a primary coolant to be classified as Class 3. This is a recommendation but not mandatory in order to consider an effect of 24Na produced by neutron irradiation in the core.
V- DIFFERENCEIN SYSTEM
......
APPROPRIATECLASSIFICATIONFOR FBR
[
COMPOSITION
COMPONENTS
IIIGE TEMPERATURE(HIGH
I
THERMALSTRESSES)
i [
DESIGN BY ANAI,YSIS AT
INTRODUCTIONOF ELEVATEDTEHPERATURE
CREEPTEMPERATURES
STRUCTURAl,DESIGNGUIDE
CONSIDERATIONOF THERHAL T
(ADDITIONAL) "DESIGN BY ANALYSIS"
I
STRESSES
]
FOR CLASS 3 COMPONENTS
i
CHARACTERISTIC FEATURES
i
OF ESSENTIAl, COMPONENTS
NON AX1SYMMETRIC
J
STRUCTURES (G/V)
[
NEGLIGIBLE POSSIBILITY
EXCLUSIONOF REOU1REMENTFOR SAEETY
OF OVERPRESSIJRE
VALVESTO SODIUMCIRCUIT
I USE OF SODIUMAS COOLANT
....
L USE OF AUSTENETICSTAIN. E
ONLYTENSILE TESTS FOR SURVEILLANCE
[
TEST
i
LESSSTEEL
!
k LOW PRESSURE
APPLICABILITY OF PNEUMATICTEST
Fig. 2. Characteristic features of Structural Design Code for Monju.
286
K. /ida et aL / Construction codes for Prototype f B R Monlu
3.2. Struetural design at creep temperatures C C P B R provides methods of design at elevated temperatures where the creep effect plays an i m p o r t a n t role in stress analysis and failure evaluation. C C P B R denotes to make this elevated temperature design with Class 1 c o m p o n e n t s at temperatures exceeding 3 7 5 ° C for ferritic steels and 425 ° C for austenetic steels/alloys. Formal Notice No. 501 for L W R s is available for Class 1 c o m p o n e n t s of M o n j u below these temperature limits. Values of allowable limits are given in C C P B R above the temperature limits with some newly developed simplified inelastic analysis methods applicable in creep range. Detailed interpretations will be given in the other p r e s e n t a t i o n with respect to the above matters. 3.3. Thermal strea:s / f a i l u r e analysis in ('lass 3 / 4 C C P B R denotes a use of " D e s i g n by F o r m u l a " in Class 3 / 4 c o m p o n e n t s principally, which is defined in Notice No. 501 for LWRs. However Class 3 / 4 components of M o n j u may suffer f r e q u e n t / s e v e r e thermal transients when they contain a sodium coolant, and Notice No. 501 does not take into account this condition. C C P B R denotes an additional requirement of following the evaluation method for Class 1 c o m p o n e n t s when c o m p o n e n t s of Class 3 / 4 suffer significant thermal stresses. This supplemental r e q u i r e m e n t is imposed to Class 3 / 4 c o m p o n e n t s containing sodium coolant at metal temperatures exceeding a f o r e m e n t i o n e d temperature limits.
exclusion of the requirement for safety valvc., in ~h~primary loop of Monju. In the secondary loop, a ,~ignificant pressure mcrcast.: may occur if sodium leaks from a heat transfer tube of a steam generator. However this event is estimated to occur with a very small probability and we can restricl a stress prodt.ced by this event within an allowable limit for Operating C o n d i t i o n II1 or IV, m the design p r a c rices. This implies that safety valves are not necessary t,~ be installed in the secondary loop. Ioo. 3.6. Surueillance test A surveillance test is imposed by Notice No. 501 for L W R , to be conducted with tensile and impact specimens irradiated in pile. A purpose of this surveillance test is to check a d u c t i l e - b r i t t l e transition of the reactor vessel material due to neutron irradiation and this check is necessary when the vessel material is ferritic steel. The reactor vessel of Monju is made of 304 stainless steel which shows no irradiation brittleness. However, a reduction of fracture ductility is observed with 304 stainless steel in high neutron irradiation condition. Also, the irradiation affects the creep ductility. In CCPBR, it is denoted that there is no need to check neutron irradiated d u c t i l e - b r i t t l e transition. C C P B R requires a surveillance test only with tensile tests. The purpose of this requirement is to supplement data to check the reduction of p l a s t i c / c r e e p ductility due to irradiation. Basic data regarding this p h e n o m e n o n must be obtained by off-site irradiation tests, 3. 7. l'ressure tests
3.4. Detailed structural design for Class 3 components C C P B R provides an optional m e t h o d for the structural design of Class 3 c o m p o n e n t s , that is, Class 3 c o m p o n e n t s may be designed using the m e t h o d given for Class 1 components, if it is desired. This option is applied to a guard vessel which is not always axisymmetric where a design formula is not given in CCPBR. 3.5. Safety t~alues in sodium circuits The installation of safety valves is imposed by Ordin a n c e No. 62 for L W R primary coolant loops, which intends to relieve a high transient pressure of any ass u m e d origin. However in the case of L M F B R , the internal pressure in the sodium coolant circuits is very low and there occurs no such event of producing significant pressure transients. This situation results in the
Pressure tests of L W R s in P S I / I S I are made with a hydrostatic test for Class 1 components. The significance of this pressure test is well understood in L W R plants as their operating pressure is rather high. But in the case of Monju, the pressure test is rather meaningless as the internal pressure is not a d o m i n a n t factor governing Class 1 s t r u c t u r e / c o m p o n e n t s . Also, a use ot a hydraulic pressure needs a very careful cleaning of the c o m p o n e n t s to delete hydro dregs in there. In C C P B R , a use of a p n e u m a t i c test is denoted for the pressure test in PSI of Class 1 c o m p o n e n t s
4. Technical standards for weldment Welding procedures for Monju c o m p o n e n t s can be said to be similar to those for L W R c o m p o n e n t s even though operating conditions may be different. There-
K. lida et al. / Construction codes for Prototype FBR Monju fore slight modifications are required on technical s t a n d a r d s for weldment, as shown in fig. 3. 4.1. Classification of components The same modifications as in the structural design code for M o n j u are required to classify structural comp o n e n t s in the technical standards for weldment for Monju. In addition special care should be taken for the guard vessel, which is classified as a Class 3 atmospheric storage tank. It was considered to be better that the guard vessel should be welded by applying the same rules as those for ordinary Class 3 vessels because of its importance in plant safety design. 4.2. Additional requirement for volumetric" or surface examinations to welds or weld preparation surfaces In the case of LWRs, welds in Class 1 pipings, p u m p s and valves with a nominal diameter of 2 in. or less, and welds for a t t a c h m e n t s to Class 1 c o m p o n e n t s are excluded from legal volumetric examinations. In the case of FBR, where thermal stresses may be
287
high, these welds should be taken care of to achieve structural integrity. From this point of view, such welds are requested to be examined volumetrically where thermal loadings cause significant stresses. Similar consideration is paid to surface examination of weld edge preparation surfaces. Weld edge preparation surfaces of a nominal thickness of 1 in. or greater should be examined where thermal loadings cause significant stresses. 4.3. Additional restriction on design of welds" A N R E Notice No. 9683 for L W R s provides that partial-penetration welds can be allowed almost for a nominal diameter of 2 in. or less and that metal backing strips are not necessarily removed after welding for Class 3 and Class 4 components. There are no restrictions on diameter for full-penetration corner welds in A N R E Notice No. 9683 for LWRs. It was noticed that such designs of welds has a tendency to locate welds near the structural discontinuities where noticable stress concentration may occur u n d e r thermal transients. Accordingly, it was concluded that such welds should
DIFFERENCE IN SYSTEM
APPROPRIATECLASSIFICATIONFOR FBR
COMPOSITION
COMPONENTS (MODIFICATIONON G/V)
•
i
! HIGH TEMPERATURE(HIGH
CONSIDERATIONOF THERHAL
THERMALSTRESSES)
STRESSES
I
ADDITIONALREOUIREMENTSFOR VOI,UHETRIC/SURFACE EXAMINATIONSTO WELDS/ WELDEDGEPREPARATIONSURFACE ADDITIONALRESTRICTIONON WELD
CHARACTERISTICFEATURES
DESIGN (PARTIALPENETRATIONWELDS,
OF ESSENTIALCOMPONENTS
CORNERWELDS~ METALBACKINGSTRIPS)
F INTEGRALSTRUCTUREBY
....
WELDHENT -
USE OF SODIUM AS COOLANT
PENETRATIONBUTTJOINTS WELDEDFROH A SINGLE SIDE
/ ---~
L LOWPRESSURE
EXPANSIONOF APPLICABILITYOF FULL
•
APPLICABILITYOF PNEUHATICTEST
l
ALTERNATIVEHETHODUSINGNON DESTRUC
]
TIVE EXAMINATIONS Fig. 3. Characteristic features of technical standard for weldment for Monju.
288
K. lida et al. / Conxtruction codes for ProtoO'pe FBR MoJ~ju
JN
,]',
. . . . . [~ii PRESSIIRE [FS[
i i
Y
,
I
RT OR ROORESSlVE PT
A S E ET,O L . OLASS- OR",,
STAINLESSSTEEL AT P R E S E N T :
~
I I
. . . .
~-
'OUTOF OBJEOTIVEOF THE PRESSURETEST
Fig. 4. Alternative method of pressure test by non destructive examination. be restricted more rigorously that in the case of LWRs. Restriction to allow partial p e n e t r a t i o n welds was reduced from 2 in. or less to 1 in. or less in nominal diameters, full-penetration corner welds were to be allowed for a n o m i n a l diameter 4 in. or less, a n d removal of metal backing strips was required, where thermal loadings cause significant stresses.
w h e n the m a x i m u m operating temperature is less than 1 0 0 ° C and the m a x i m u m operating pressure is less than 20 k g / c m 2. The definition of " m o d e r a t e energy fluid system" is generalized in the technical s t a n d a r d for w e l d m e n t for M o n j u such that the m a x i m u m temperature is less than the boiling point of the internal fluid.
4. 5. Pressure test 4.4. Expansion of applicability of full penetration butt ,joints welded from a single side Applicability of full p e n e t r a t i o n butt joints, welded from a single side such as using T I G weld for the first layer, is restricted for Class 1 vessels of an inner diameter of 600 m m or less by A N R E Notice No. 9683. The Class 1 c o m p o n e n t s of M o n j u are fabricated almost without mechanical joints on the sodium boundary. Application of full p e n e t r a t i o n butt j o i n t s from a single side using T I G weld for the first layer, which can be considered to have sufficient good quality in these days, is required to the portions of the I H X s where the inner diameter exceeds 600 mm. Consequently it was concluded that such j o i n t s could be allowed for Class 1 vessels even if their inner d i a m e t e r exceeds 600 mm. In addition, applicability of such j o i n t s for Class 3 or Class 4 c o m p o n e n t s of " m o d e r a t e energy fluid system" is provided by A N R E Notice No. 9683 assuming water or steam as the internal fluid. They can be applied to Class 3 or Class 4 c o m p o n e n t s
Modifications necessary for the rules of the pressure test are same as in the structural design code for Monju. In addition, a reasonable alternative m e t h o d by non-destructive e x a m i n a t i o n as shown on fig. 4 was recommended.
5. Conclusion A C o n s t r u c t i o n Code a n d Standards for a Prototype Fast Breeder Reactor were developed. They were based on those for LWRs, and several modifications were required by taking into account characteristic features of L M F B R structures and environment. They were proposed to the regulatory authorities, a n d after extensive deliberation by technical advisers, Science a n d Technology A u t h o r i t y published a set of c o d e s / s t a n d a r d s / g u i d e s for Fast Breeder Reactors. The C o n s t r u c t i o n Code a n d Standards that are presented in this paper were incorporated in them.
CONSTRUCTION
283
CODES DEVELOPED FOR PROTOTYPE
FBR MONJU
K u n i h i r o I I D A 1, Yasuhide A S A D A 1, K u n i o O K A B A Y A S H I 2 a n d Takashi N A G A T A 2 l Facul(vof Engineering, The Universi O' of ToA3o, 7-3-1 ttongo, Bunkyo-ku, Tokyo ll3, Japan -~Power Reactor and Nuclear Fuel Development Corporation, Sankaido Bldg. 1-9-13 Akasaka, Minato-ku. Tokyo 107, Japan
Received May 1986
The paper describes the regulation system for LWRs in Japan and discusses some typical points which are attained in the newly developed CCPBR (Construction Code for Prototype Breeder Reactor). Revised welding standards are briefly introduced for the Monju construction.
1. Introduction The safety evaluation has been accomplished in May 1983 with Japanese Prototype Breeder Reactor "Monju" which has been established after around two decades research and development activities at Power Reactor and Nuclear Fuel Development Corporation (PNC). Following this evaluation, Monju is now proceeding with an application for Construction permit to the government. The Japanese Government has been preparing necessary codes, standards and guides for regulating construction and operations of nuclear power plant stations. However, these codes/standards/guides are, almost in all the cases, applicable to Light Water Reactors. Some modifications or revisions are necessary for their application to Monju. The main part of these revisions is in the area of construction, examination and testing of structures/ components and welding of Monju. For this purpose, PNC has proposed "Elevated Temperature Structural Design Guide for Class 1 Components of Prototype Fast Breeder Reactor" ("ETSDG") which enables design of structures/components in creep range with some additional considerations taking into account some other characteristic features of LMFBR structures, that is, being subjected to high/frequent thermal transients, use of sodium coolant, and thin walled structure. The Construction Code for Prototype Breeder Reactor (CCPBR) has been developed in 1983-84 to give guidance for structural design, examinations and testing of Monju on a basis of ETSDG. The present paper introduces, at first, the regulation system for LWRs in
Japan and then discusses some typical points which are attained in the newly developed CCPBR. Revised welding standards are also briefly introduced for the Monju construction.
2. Current regulation system for LWRs in Japan Agency of National Resouces and Energy (ANRE) of the Ministry of International Trade and Industries (MITI) is the responsible organization in Japan for construction and operation of nuclear power stations in Japan. ANRE has prepared codes, standards and guides for regulating the structural design, examination and testing for construction and operation of nuclear power plants. The regulation is principally based upon Electric Utility Industry Law (EUIL) which gives three articles with respect to structural design of nuclear power plants, that is, Art. 41 (Construction Permission), Art. 46 (Examination of Welding) and Art. 48 (Maintenance). Some ordinances are prepared on a basis of these articles. MITI Ordinance No. 62 (1965) "Technical Standard for Facilities of Nuclear Power Plant" gives Art. 5 (Resistance against Earthquake), Art. 9 (Materials and Structural Design), Art. 10 (Safety Valves), Art. 11 (Pressure Test) and Art. 12 (Surveillance Test). Detailed descriptions of these articles are given in Notice and Regulartory Guide. MITI Ordinance No. 81 gives "Technical Standard for Weldment of Structures for Electricity". ANRE Ordinance No. 9683 (1975) gives "Notification as to Licensing of Welding Method". Details regarding structural design, examinations and
0 0 2 9 - 5 4 9 3 / 8 7 / $ 0 3 . 5 0 © Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d Physics Publishing Division)
284
K. hda et al. / Construction codes fi~r Prototype FBR Mo±uu
testing are given in MITI Notice No. 501 (1980) "Technical Standard for Structural Design of Facilities of Nuclear Power Stations" which is constructed upon a same concept as that of A S M E Boiler & Pressure Vessel Code Section III, although some revisions are included in order to adjust the descriptions to Japanese social circumstances. An aseismic design code is given by MITI Regulatory Guide "Standard for Design of
l
[
Nuclear Power Plants against Earthquake". m detail. The relation between these codes, standards and guides are schematically given in fig. I with regard It, articles relating structural design, examinations and testing. An important restriction m development of C C P B R is to incorporate ETSDG with current regulation systems by conserving the current c o d e / s t a n d a r d system for LWRs.
ELECTRIC UTILITY INDIJSTRY[,AN
r
!
&rt.41
Construction Permission
Art.46
Examination of Weldment
~ Art.48
]
Maintenance
MITI ORDINANCENO. 62 (1965) Technical Standard for Facilities of Nuclear
MITI: Ministry of I n terna t i ona 1 Trade
Power Plant Art. 5 Resistance against Earthquake
and Industry
Mr. 9 Material and Structural Design Art. lO Safety Valves
E
Art.t1
Pressure Test
iI
Art.12
Surveillance Test I
MITI NOTICENO. 501 (1980)
J
Technical Standard for Structural Design, etc. of Facilities of Nuclear Power Plant MITI REGULATORYGUIOE Standard for Design against Earthquake of Nuclear Power Plant MITI ORDINANCENO. 81 (1970) Technical Standard for Weldment of Structures for Electricity ANRE NOTICENO. 9683 (1975) Notification as to Licensing of Welding Methods
ANRE: Agency of Natural Resources and Energy
Fig. 1. Construction codes for LWRs under the Electric Utility Industry Law.
K. lida et al. / Construction codes for Proto(vpe FBR Monju
3. Construction Code for Prototype Breeder Reactor (CCPBR) In this section, topics of C C P B R are briefly introduced with their fundamental concepts a n d / o r backgrounds. C C P B R resulted from an incorporation of MITI Notice No. 501 with E T S D G which has been developed at PNC after around two decade research activities. E T S D G will be introduced in this seminar by the same authors as this one. E T S D G is established upon a same basic concept to that of A S M E Code Case N-47. However, E T S D G includes some advanced procedures regarding creep-fatigue evaluation, elastic follow-up evaluation and others on a basis of elastic analyses. Fig. 2 shows a conceptual framework of CCPBR.
3. I. Classification o f components In CCPBR, a classification of c o m p o n e n t s is at first denoted. This classification is made principally upon a degree of importance of the role of c o m p o n e n t s in plant safety design. Therefore the classification of components in Monju is a little bit different from that in L W R s according to the difference of system safety of L M F B R and LWR.
285
A first revision of classification is made with some concepts in a primary coolant loop. A shield plug of a reactor vessel, a cover gas system and a overflow system in a primary coolant loop are categorized as Class 3. On the other hand in LWRs, most of c o m p o n e n t s belonging to the primary loop are categorized as Class 1. This revision is made on a base of a heat removal from the core of L M F B R in which pressurization of a coolant (sodium) is not necessary for this purpose but a necessary point at this viewpoint is to maintain a level of a coolant in the core. That is, a function of the heat removal from the core is satisfied as long as a level of the coolant is kept and is not affected from a depressurization of it, in case of LMFBR. Please note that Class 3 means in Japanese system a class corresponding to Class 2 in A S M E Code. Class 2 in Japanese system is used for categorizing containment vessels which belong to Class MC in A S M E Code. Another revision of classification is made with all c o m p o n e n t s other than Class 1 which contain a primary coolant to be classified as Class 3. This is a recommendation but not mandatory in order to consider an effect of 24Na produced by neutron irradiation in the core.
V- DIFFERENCEIN SYSTEM
......
APPROPRIATECLASSIFICATIONFOR FBR
[
COMPOSITION
COMPONENTS
IIIGE TEMPERATURE(HIGH
I
THERMALSTRESSES)
i [
DESIGN BY ANAI,YSIS AT
INTRODUCTIONOF ELEVATEDTEHPERATURE
CREEPTEMPERATURES
STRUCTURAl,DESIGNGUIDE
CONSIDERATIONOF THERHAL T
(ADDITIONAL) "DESIGN BY ANALYSIS"
I
STRESSES
]
FOR CLASS 3 COMPONENTS
i
CHARACTERISTIC FEATURES
i
OF ESSENTIAl, COMPONENTS
NON AX1SYMMETRIC
J
STRUCTURES (G/V)
[
NEGLIGIBLE POSSIBILITY
EXCLUSIONOF REOU1REMENTFOR SAEETY
OF OVERPRESSIJRE
VALVESTO SODIUMCIRCUIT
I USE OF SODIUMAS COOLANT
....
L USE OF AUSTENETICSTAIN. E
ONLYTENSILE TESTS FOR SURVEILLANCE
[
TEST
i
LESSSTEEL
!
k LOW PRESSURE
APPLICABILITY OF PNEUMATICTEST
Fig. 2. Characteristic features of Structural Design Code for Monju.
286
K. /ida et aL / Construction codes for Prototype f B R Monlu
3.2. Struetural design at creep temperatures C C P B R provides methods of design at elevated temperatures where the creep effect plays an i m p o r t a n t role in stress analysis and failure evaluation. C C P B R denotes to make this elevated temperature design with Class 1 c o m p o n e n t s at temperatures exceeding 3 7 5 ° C for ferritic steels and 425 ° C for austenetic steels/alloys. Formal Notice No. 501 for L W R s is available for Class 1 c o m p o n e n t s of M o n j u below these temperature limits. Values of allowable limits are given in C C P B R above the temperature limits with some newly developed simplified inelastic analysis methods applicable in creep range. Detailed interpretations will be given in the other p r e s e n t a t i o n with respect to the above matters. 3.3. Thermal strea:s / f a i l u r e analysis in ('lass 3 / 4 C C P B R denotes a use of " D e s i g n by F o r m u l a " in Class 3 / 4 c o m p o n e n t s principally, which is defined in Notice No. 501 for LWRs. However Class 3 / 4 components of M o n j u may suffer f r e q u e n t / s e v e r e thermal transients when they contain a sodium coolant, and Notice No. 501 does not take into account this condition. C C P B R denotes an additional requirement of following the evaluation method for Class 1 c o m p o n e n t s when c o m p o n e n t s of Class 3 / 4 suffer significant thermal stresses. This supplemental r e q u i r e m e n t is imposed to Class 3 / 4 c o m p o n e n t s containing sodium coolant at metal temperatures exceeding a f o r e m e n t i o n e d temperature limits.
exclusion of the requirement for safety valvc., in ~h~primary loop of Monju. In the secondary loop, a ,~ignificant pressure mcrcast.: may occur if sodium leaks from a heat transfer tube of a steam generator. However this event is estimated to occur with a very small probability and we can restricl a stress prodt.ced by this event within an allowable limit for Operating C o n d i t i o n II1 or IV, m the design p r a c rices. This implies that safety valves are not necessary t,~ be installed in the secondary loop. Ioo. 3.6. Surueillance test A surveillance test is imposed by Notice No. 501 for L W R , to be conducted with tensile and impact specimens irradiated in pile. A purpose of this surveillance test is to check a d u c t i l e - b r i t t l e transition of the reactor vessel material due to neutron irradiation and this check is necessary when the vessel material is ferritic steel. The reactor vessel of Monju is made of 304 stainless steel which shows no irradiation brittleness. However, a reduction of fracture ductility is observed with 304 stainless steel in high neutron irradiation condition. Also, the irradiation affects the creep ductility. In CCPBR, it is denoted that there is no need to check neutron irradiated d u c t i l e - b r i t t l e transition. C C P B R requires a surveillance test only with tensile tests. The purpose of this requirement is to supplement data to check the reduction of p l a s t i c / c r e e p ductility due to irradiation. Basic data regarding this p h e n o m e n o n must be obtained by off-site irradiation tests, 3. 7. l'ressure tests
3.4. Detailed structural design for Class 3 components C C P B R provides an optional m e t h o d for the structural design of Class 3 c o m p o n e n t s , that is, Class 3 c o m p o n e n t s may be designed using the m e t h o d given for Class 1 components, if it is desired. This option is applied to a guard vessel which is not always axisymmetric where a design formula is not given in CCPBR. 3.5. Safety t~alues in sodium circuits The installation of safety valves is imposed by Ordin a n c e No. 62 for L W R primary coolant loops, which intends to relieve a high transient pressure of any ass u m e d origin. However in the case of L M F B R , the internal pressure in the sodium coolant circuits is very low and there occurs no such event of producing significant pressure transients. This situation results in the
Pressure tests of L W R s in P S I / I S I are made with a hydrostatic test for Class 1 components. The significance of this pressure test is well understood in L W R plants as their operating pressure is rather high. But in the case of Monju, the pressure test is rather meaningless as the internal pressure is not a d o m i n a n t factor governing Class 1 s t r u c t u r e / c o m p o n e n t s . Also, a use ot a hydraulic pressure needs a very careful cleaning of the c o m p o n e n t s to delete hydro dregs in there. In C C P B R , a use of a p n e u m a t i c test is denoted for the pressure test in PSI of Class 1 c o m p o n e n t s
4. Technical standards for weldment Welding procedures for Monju c o m p o n e n t s can be said to be similar to those for L W R c o m p o n e n t s even though operating conditions may be different. There-
K. lida et al. / Construction codes for Prototype FBR Monju fore slight modifications are required on technical s t a n d a r d s for weldment, as shown in fig. 3. 4.1. Classification of components The same modifications as in the structural design code for M o n j u are required to classify structural comp o n e n t s in the technical standards for weldment for Monju. In addition special care should be taken for the guard vessel, which is classified as a Class 3 atmospheric storage tank. It was considered to be better that the guard vessel should be welded by applying the same rules as those for ordinary Class 3 vessels because of its importance in plant safety design. 4.2. Additional requirement for volumetric" or surface examinations to welds or weld preparation surfaces In the case of LWRs, welds in Class 1 pipings, p u m p s and valves with a nominal diameter of 2 in. or less, and welds for a t t a c h m e n t s to Class 1 c o m p o n e n t s are excluded from legal volumetric examinations. In the case of FBR, where thermal stresses may be
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high, these welds should be taken care of to achieve structural integrity. From this point of view, such welds are requested to be examined volumetrically where thermal loadings cause significant stresses. Similar consideration is paid to surface examination of weld edge preparation surfaces. Weld edge preparation surfaces of a nominal thickness of 1 in. or greater should be examined where thermal loadings cause significant stresses. 4.3. Additional restriction on design of welds" A N R E Notice No. 9683 for L W R s provides that partial-penetration welds can be allowed almost for a nominal diameter of 2 in. or less and that metal backing strips are not necessarily removed after welding for Class 3 and Class 4 components. There are no restrictions on diameter for full-penetration corner welds in A N R E Notice No. 9683 for LWRs. It was noticed that such designs of welds has a tendency to locate welds near the structural discontinuities where noticable stress concentration may occur u n d e r thermal transients. Accordingly, it was concluded that such welds should
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K. lida et al. / Conxtruction codes for ProtoO'pe FBR MoJ~ju
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w h e n the m a x i m u m operating temperature is less than 1 0 0 ° C and the m a x i m u m operating pressure is less than 20 k g / c m 2. The definition of " m o d e r a t e energy fluid system" is generalized in the technical s t a n d a r d for w e l d m e n t for M o n j u such that the m a x i m u m temperature is less than the boiling point of the internal fluid.
4. 5. Pressure test 4.4. Expansion of applicability of full penetration butt ,joints welded from a single side Applicability of full p e n e t r a t i o n butt joints, welded from a single side such as using T I G weld for the first layer, is restricted for Class 1 vessels of an inner diameter of 600 m m or less by A N R E Notice No. 9683. The Class 1 c o m p o n e n t s of M o n j u are fabricated almost without mechanical joints on the sodium boundary. Application of full p e n e t r a t i o n butt j o i n t s from a single side using T I G weld for the first layer, which can be considered to have sufficient good quality in these days, is required to the portions of the I H X s where the inner diameter exceeds 600 mm. Consequently it was concluded that such j o i n t s could be allowed for Class 1 vessels even if their inner d i a m e t e r exceeds 600 mm. In addition, applicability of such j o i n t s for Class 3 or Class 4 c o m p o n e n t s of " m o d e r a t e energy fluid system" is provided by A N R E Notice No. 9683 assuming water or steam as the internal fluid. They can be applied to Class 3 or Class 4 c o m p o n e n t s
Modifications necessary for the rules of the pressure test are same as in the structural design code for Monju. In addition, a reasonable alternative m e t h o d by non-destructive e x a m i n a t i o n as shown on fig. 4 was recommended.
5. Conclusion A C o n s t r u c t i o n Code a n d Standards for a Prototype Fast Breeder Reactor were developed. They were based on those for LWRs, and several modifications were required by taking into account characteristic features of L M F B R structures and environment. They were proposed to the regulatory authorities, a n d after extensive deliberation by technical advisers, Science a n d Technology A u t h o r i t y published a set of c o d e s / s t a n d a r d s / g u i d e s for Fast Breeder Reactors. The C o n s t r u c t i o n Code a n d Standards that are presented in this paper were incorporated in them.